2191 R0 XI.M7

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XI.M7 BWR STRESS CORROSION CRACKING

Program Description

The program to manage intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) coolant pressure boundary piping made of stainless steel (SS) and nickel-based alloy components is delineated in NUREG-0313, Rev. 2, and US Nuclear Regulatory Commission (US NRC) Generic Letter (GL) 88-01 and its Supplement 1. The material includes base metal and welds. The comprehensive program outlined in NUREG-0313, Rev. 2 and US NRC GL 88-01 describes improvements that, in combination, will reduce the susceptibility to IGSCC. The elements to cause IGSCC consist of a susceptible–material, a significant tensile stress, and an aggressive environment. Sensitization of nonstabilized austenitic SSs containing greater than 0.035 weight percent carbon involves precipitation of chromium carbides at the grain boundaries during certain fabrication or welding processes. The formation of carbides creates a chromium-depleted region that, in certain environments, is susceptible to stress corrosion cracking (SCC). Residual tensile stresses are introduced from fabrication processes, such as welding, cold work, surface grinding, and forming. High levels of dissolved oxygen or aggressive contaminants, such as sulfates or chlorides, accelerate the SCC processes. The program includes (a) preventive measures to mitigate IGSCC and (b) inspection and flaw evaluation to monitor IGSCC and its effects. The staff-approved Boiling Water Reactor Vessel and Internals Project (BWRVIP)-75-A report allows for modifications to the inspection extent and schedule described in the US NRC GL 88-01 program.

Evaluation and Technical Basis

1. Scope of Program: The program focuses on (a) managing and implementing countermeasures to mitigate IGSCC and (b) performing ISI to monitor IGSCC and its effects on the intended function of BWR piping components within the scope of license renewal. The program is applicable to all BWR piping and piping welds made of austenitic–SS and nickel alloy that are 4 inches or larger in nominal diameter containing reactor coolant at a temperature above 93 °C (Celsius) [200 °F (Fahrenheit)] during power operation, regardless of code classification. The program also applies to pump casings, valve bodies, and reactor vessel attachments and appurtenances, such as head spray and vent components. Control rod drive return line nozzle caps and associated welds (previously addressed in Generic Aging Lessons Learned (GALL) Report, Revision 2, AMP XI.M6, “BWR Control Rod Drive Return Line Nozzle”) may be included in the scope of the program. NUREG-0313, Rev. 2 and US NRC GL 88-01, respectively, describe the technical basis and staff guidance regarding mitigation of IGSCC in BWRs. Attachment A of US NRC GL 88-01 delineates the staff-approved positions regarding materials, processes, water chemistry, weld overlay reinforcement, partial replacement, stress improvement of cracked welds, clamping devices, crack characterization and repair criteria, inspection methods and personnel, inspection schedules, sample expansion, leakage detection, and reporting requirements.

2. Preventive Actions: The BWR SCC program is primarily a condition monitoring program which also relies on countermeasures. Maintaining high water purity reduces susceptibility to SCC or IGSCC. Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry program. The program description, evaluation and technical basis of water chemistry are addressed through implementation of the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report AMP XI.M2, “Water Chemistry.” In addition, NUREG-0313, Rev. 2 and GL 88-01 delineate the guidance for selection of resistant materials and processes that provide resistance to IGSCC such as solution heat treatment and stress improvement processes.

3. Parameters Monitored or Inspected: The program detects and sizes cracks and detects leakage by using the examination and inspection guidelines delineated in NUREG-0313, Rev. 2 and US NRC GL 88-01.

4. Detection of Aging Effects: The extent, method, and schedule of the inspection and test techniques delineated in US NRC GL 88-01 are designed to maintain structural integrity, to detect and mitigate degradation, and to repair or replace components before the loss of intended function of the component. Modifications to the extent and schedule of inspection in US NRC GL 88-01 are allowed in accordance with the inspection guidance in approved BWRVIP-75-A. The potential for stagnant flow conditions such as dead legs is considered when selecting inspection locations. The program identifies these locations. Prior to crediting hydrogen water chemistry to modify extent and frequency of inspections in accordance with BWRVIP-75-A, the applicant should meet conditions described in the staff’s safety evaluations regarding BWRVIP-62-A(Archived). The program uses volumetric examinations to detect IGSCC. Inspection can reveal cracking and leakage of coolant. The extent and frequency of inspection recommended by the program are based on the condition of each weld (e.g., whether the weldments were made from IGSCC-resistant material, whether a stress improvement process was applied to a weldment to reduce residual stresses, and how the weld was repaired, if it had been cracked).

5. Monitoring and Trending: The extent and schedule for inspection, in accordance with the recommendations of US NRC GL 88-01 or approved BWRVIP-75-A guidelines, provide timely detection of cracks and leakage of coolant. Indications of cracking are evaluated and trended in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, IWA-3000.

Applicable and approved BWRVIP-14-A, BWRVIP-59-A, BWRVIP-60-A, and BWRVIP-62-A(Archived) reports provide guidelines for evaluation of crack growth in SSs, nickel alloys, and low-alloy steels. An applicant may use BWRVIP-61 guidelines for BWR vessel and internals induction heating stress improvement effectiveness on crack growth in operating plants.

6. Acceptance Criteria: Any cracking is evaluated in accordance with ASME Code, Section XI, IWA-3000 by comparing inspection results with the acceptance standards of ASME Code, Section XI, IWB-3000, IWC-3000 and IWD-3000 for Class 1, 2 and 3 components, respectively.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicant’s corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, “Corrective Action,” of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this aging management program (AMP) for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

The guidance for weld overlay repair and stress improvement or replacement is provided in US NRC GL 88-01. Corrective action is performed in accordance with IWA-4000.

8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

10. Operating Experience: Intergranular SCC has occurred in small- and large-diameter BWR piping made of austenitic–SS and nickel-base alloys. Cracking has occurred in recirculation, core spray, residual heat removal, control rod drive return line penetrations, and reactor water cleanup system piping welds (US NRC GL 88-01, US NRC Information Notices [INs] 82-39 and 84-41, and 04-08). The comprehensive program outlined in US NRC GL 88-01, NUREG-0313, Rev. 2, and in the staff-approved BWRVIP-75-A report addresses mitigating measures for SCC or IGSCC (e.g., susceptible material, significant tensile stress, and an aggressive environment). The GL 88-01 program, with or without the modifications allowed by the staff-approved BWRVIP-75-A report, has been effective in managing IGSCC in BWR reactor coolant pressure-retaining components and will adequately manage IGSCC degradation.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.55a, “Codes and Standards.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” New York, New York: The American Society of Mechanical Engineers. 2008.

_____. ASME Code Case N-504-4, “Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping.” Section XI, Division 1. New York, New York: American Society of Mechanical Engineers. July 2006.

EPRI. BWRVIP-14-A (EPRI 1016569), “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals.” Palo Alto, California: Electric Power Research Institute. September 2008.

_____. BWRVIP-59-A, (EPRI 1014874), “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Nickel-Base Austenitic Alloys in RPV Internals.” Palo Alto, California: Electric Power Research Institute. May 2007.

_____. BWRVIP-60-A (EPRI 108871), “BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment.” Palo Alto, California: Electric Power Research Institute. June 2003.

_____. BWRVIP-61 (EPRI 112076), “BWR Vessel and Internals Induction Heating Stress Improvement Effectiveness on Crack Growth in Operating Reactors.” Palo Alto, California: Electric Power Research Institute. January 1999.

_____. BWRVIP-62-A (EPRI-1021006), “BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection.”(Archived) Palo Alto, California: Electric Power Research Institute. November 2010.

_____. BWRVIP-75-A (EPRI 1012621), “BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (NUREG–0313).” Palo Alto, California: Electric Power Research Institute. October 2005.

US NRC Generic Letter 88-01, US NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, U.S. Nuclear Regulatory Commission, January 25, 1988; Supplement 1, February 4, 1992.

_____. Information Notice 04-08, “Reactor Coolant Pressure Boundary Leakage Attributable to Propagation of Cracking in Reactor Vessel Nozzle Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. April 2004.

_____. Information Notice 82-39, “Service Degradation of Thick Wall Stainless Steel Recirculation System Piping at a BWR Plant.” Washington, DC: U.S. Nuclear Regulatory Commission. September 1982.

_____. Information Notice 84-41, “IGSCC in BWR Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. June 1984.

_____. NUREG–0313, “Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping.” Revision 2. Washington DC: U.S. Nuclear Regulatory Commission. 1988.

Boiling Water Reactor

United States of America

Nuclear Regulatory Commission - The US federal government agency that regulates nuclear power plants

Stress Corrosion Cracking

Boiling Water Reactor Vessel Internals Program - The BWRVIP is an industry program for General Electric boiling water reactors that has developed procedures for performing inspections of reactor vessel internals.

American Society of Mechanical Engineers

Quality Assurance

Aging Management Program

Electric Power Research Institute